scholarly journals Analysis of the beryllium stability under standard and critical operation in a fusion reactor

2021 ◽  
Vol 5 (4) ◽  
pp. 188-197
Author(s):  
I. A. Sokolov ◽  
M. K. Skakov ◽  
A. Zh. Miniyazov ◽  
B. T. Aubakirov ◽  
T. R. Tulenbergenov ◽  
...  
Keyword(s):  
Heat Flux ◽  
Partial Melting ◽  
Fusion Reactor ◽  
Structural Phase ◽  
Phase Changes ◽  
First Wall ◽  
Near Wall

The paper provides data on the peculiarity of change in the structure, structural phase changes and destructions in beryllium resulting from interaction with a near-wall plasma of fusion facilities. Beryllium resistance under conditions of ITER operation was evaluated, which considers factors leading to possible partial melting and erosion of panels of the ITER first wall. It presents the modelling of a heat s distribution in element (”finger”) of the first wall at ”normal” and ”increased” heat flux of the ITER operation.

Alignment of small He-Bubbles during in situ deformation of Al-foils

1978 ◽  
Vol 36 (1) ◽  
pp. 396-397
Author(s):  
E. Ruedl ◽  
P. Schiller
Keyword(s):  
Fusion Reactor ◽  
Matrix Material ◽  
Fusion Reactors ◽  
First Wall ◽  
Potential Matrix ◽  
In Situ Deformation ◽  
Α Particles ◽  
Metal Aluminium

The low Z metal aluminium is a potential matrix material for the first wall in fusion reactors. A drawback in the application of A1 is the rel= atively high amount of He produced in it under fusion reactor conditions. Knowledge about the behaviour of He during irradiation and deformation in Al, especially near the surface, is therefore important.Using the TEM we have studied Al disks of 3 mm diameter and 0.2 mm thickness, which were perforated at the centre by double jet polishing. These disks were bombarded at∽200°C to various doses with α-particles, impinging at any angle and energy up to 1.5 MeV at both surfaces. The details of the irradiations are described in Ref.1. Subsequent observation indicated that in such specimens uniformly distributed He-bubbles are formed near the surface in a layer several μm thick (Fig.1).After bombardment the disks were deformed at 20°C during observation by means of a tensile device in a Philips EM 300 microscope.

Special Features of First-Wall Heat Transfer in Liquid-Metal Fusion Reactor Blankets

1987 ◽  
Vol 12 (1) ◽  
pp. 104-113 ◽  
Author(s):  
K. Taghavi ◽  
M. S. Tillack ◽  
H. Madarame
Keyword(s):  
Heat Transfer ◽  
Liquid Metal ◽  
Fusion Reactor ◽  
Wall Heat Transfer ◽  
First Wall

Near-Wall Microlayer Evaporation Analysis and Experimental Study of Nucleate Pool Boiling on Inclined Surfaces

1998 ◽  
Vol 120 (3) ◽  
pp. 641-653 ◽  
Author(s):  
G. F. Naterer ◽  
W. Hendradjit ◽  
K. J. Ahn ◽  
J. E. S. Venart
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Heating Surface ◽  
Nucleate Boiling ◽  
Inclined Surfaces ◽  
Wide Range ◽  
Model Predictions ◽  
Microlayer Evaporation ◽  
Near Wall

Boiling heat transfer from inclined surfaces is examined and an analytical model of bubble growth and nucleate boiling is presented. The model predicts the average heat flux during nucleate boiling by considering alternating near-wall liquid and vapor periods. It expresses the heat flux in terms of the bubble departure diameter, frequency and duration of contact with the heating surface. Experiments were conducted over a wide range of upward and downward-facing surface orientations and the results were compared to model predictions. More active microlayer agitation and mixing along the surface as well as more frequent bubble sweeps along the heating surface provide the key reasons for more effective heat transfer with downward facing surfaces as compared to upward facing cases. Additional aspects of the role of surface inclination on boiling dynamics are quantified and discussed.

Updated Design and Development Route for CH HCCB TBM and its Mockup

2013 ◽  
Author(s):  
Gang Hu ◽  
Kaiming Feng ◽  
Zhou Zhao ◽  
Guoshu Zhang ◽  
Qijie Wang ◽  
...  
Keyword(s):  
Fusion Reactor ◽  
Surface Heat ◽  
Maximum Temperature ◽  
First Wall ◽  
Total Stress ◽  
Design And Development ◽  
Tritium Production ◽  
Radial Length ◽  
Development Route ◽  
Purge Gas

Chinese helium-cooled ceramics breeder test blanket module (CH HCCB TBM) is determined to be tested in ITER machine to get data for fusion reactor design and development in future. Chinese TBM is designed to occupy half of port C with 484mm in torroidal and 1660mm in poloidal. Radial length is 675mm. TBM is composed of box, 12 submodules and independent backplate. Box formed by first wall, grids and caps have 12 caivities to hold submodules. Box and submodules are supported by backplate by welding. Backplate distribute helium with flow rate 1.36kg/s to cool first wall and then part of it go out of TBM by bypass. The rest 0.77kg/s go on to cool caps and girds first and then cool submodules. Submodules with dimensions 250mm×202mm×318mm have independent cooling and purging systems connected to backplate manifold systems. In a submodule, two U-shaped structures hold breeding material Li4SiO4 pebbles. Out of the structure filled beryllium pebbles. Neutronics results show that tritium production is ∼64mg/FPD. Maximum temperature 538°C of structure material occurs in the front of first wall with surface heat flux 0.5MW/m2. Maximum total stress at first wall is 471MPa at 394°C; that in submodules is 426MPa at 400°C; that in backplate is 526MPa at 410°C, In order to explore development technologies for the TBM, a mockup with dimensions 484mm (torroidal)×592mm (poloidal)×675mm (radial) has been designed. The mockup with similar structure ignores bypass and purge gas systems. In the mockup, there’s only one submodule and the other three are replaced by submodule replacements. By discussions and investigations, development route has been decided and the mockup is being fabricated.

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